Damage Energy Cross Sections Research Articles

Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO_FUS based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils and material damage (dpa, gas production) to estimate the operational life of the steel-made structural components in the blanket and vacuum vessel (VV). In order to optimize the shielding of the coils different combinations of water and steel have been considered for the gap of the VV. The used neutron source is based on an axi-symmetric 2D fusion reaction profile for the given plasma equilibrium configuration. MCNPX has been used for transport calculations and ACAB has been used to handle gas production and damage energy cross sections.

Calculation and analysis of DPA in the main components of CSNS target station

The damage energy cross sections and displacement cross sections of W, SS316 and Al-6061, which will be used in the CSNS (China Spalllation Neutron Source) target station, were calculated with the high energy transport code MCNPX 2.5.0. The DPA (Displacement Per Atom) distributions were also calculated and anlyzed for the main components in the CSNS target station, such as the W target, the target vessel, the moderators and reflector vessels. Based on the DPA caused by neutrons and protons, the lifetime of these components were estimated. The influence of the proton beam profile on the peak value of DPA for W target and target vessel were also presented. These calculations and analyses are very important for the under going CSNS project.

The NCSU radiation damage database; proton-induced damage energy and application to radiation damage at SINQ

The NCSU radiation damage database is described. It contains proton and neutron cross sections for production of damage energy, displacements, helium, hydrogen, and heavier transmutation products. The targets in the database include 23 elements from Mg to U and eight practical alloys, but for this paper attention is focused on Al, Fe, and W. Damage energy cross sections are presented for 20–3200-MeV protons based on intranuclear cascade (INC) models (Bertini, CEM2k, and ISABEL) and for 1–2000-MeV protons based on classical Rutherford scattering and SRIM. These cross sections are used to calculate displacement production at SINQ Target 5. It is shown that the SRIM-calculated cross sections can make a significant contribution to proton-induced displacement production in Al at the entrance window and in Fe at Rods 1 and 15.

Calculation of displacement, gas, and transmutation production in stainless steel irradiated with spallation neutrons

Calculations using the high-energy transport code LAHET have been made for the production of displacements, helium gas, and transmuted atoms for stainless steel (Fe-18 wt% Cr-10 wt% Ni) irradiated with spallation neutrons at energies of 100 to 1600 MeV. The damage energy cross section increased from about 250 to 350 b keV for increasing neutron energies from 100 to 1600 MeV with a spallation spectrum average of 281 b keV. For a displacement threshold energy of 33 eV, the corresponding spectrum-average displacement cross section is 3400 b. The PKA spectrum was found to be fairly independent of the incident neutron energy, with an average damage energy of 0.25–0.30 MeV. The helium production cross section increased monotonically with increasing neutron energy with a spectrum average of 0.32 b. The maximum transmutation yield was observed near manganese ( Z = 25), corresponding to a production cross section of about 0.2 b. Relevance to fusion materials is discussed.

Resistivity recovery in high purity iron after fission- and fusion- neutron irradiation

A resistometric study of high-purity iron has been performed after irradiation at 20 K by 14 MeV neutrons in RTNS-II or by fission neutrons in Kyoto University Reactor (KUR). The annealing behavior of iron after fusion neutrons is very different from that after electron irradiation. The fractional recovery in stage I is much smaller and the recovery of vacancy type defects starts to occur at a much lower temperature with a lower activation energy than after electron iradiation. The difference between fission and fusion neutron irradiation, however, is much smaller; the isochronal curves for these two types of irradiation are essentially the same below 170 K. It is concluded that the local defect configuration for these two types of irradiation is similar. The induced resistivity by fusion neutrons is about a factor of 2 larger than by fission neutrons (fluence for E > 1 MeV). This is in accord with a rough estimation of the ratio of damage energy cross sections.

Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10 17, 6.1 × 10 16 ( E < 0.55 eV), 7.6 × 10 16 ( E > 0.1 MeV) and 4.0 × 10 16 ( E > 1 MeV) m −2 s −1. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.

Flow behavior of nickel irradiated with 15 MeV neutrons and 16 MeV protons

The yield strength of zone refined nickel that was irradiated at 300 K with T(d, n) and Be(d, n) neutrons and 16 MeV protons was found to increase by a similar amount when compared at equal fluence levels. A factor of 2.5 increase in the yield strength was determined for a 15 MeV neutron fluence of 1 × 10 22 m −2 and a 16 MeV proton fluence of 2 × 10 22 m −2 . The strength increase was associated with defect clusters or small prismatic loops and could be described by a strong barrier model for both neutron and proton irradiated nickel. Serrated plastic deformation was observed for samples irradiated with 15 MeV neutrons and 16 MeV protons to fluences higher than 3 × 10 21 m −2 . The similarity in the hardening process of 15 MeV neutron and 16 MeV proton irradiated nickel is consistent with their damage energy cross sections being equal at high primary knock-on energies (> 10 keV).

Sputtering with neutrons

Any radiation causing damage in the bulk of a material also causes sputtering, i.e. the release of atoms at the surface. For neutrons, we have to distinguish between backward, forward and isotropic sputtering, depending upon the direction of the neutron flux. The sputtering yields, defined as the mean number of atoms removed from the surface per incident neutron, are generally expected to be proportional to the specific damage energy deposited in the surface layers or the damage energy cross-section, $ ̌ σ n ( E) v( T). The cross-section, σ n ( E), refers to elastic and inelastic scattering as well as the different nucle reactions. If the primary knock-on atoms are radioactive their emission yield can be detected separately. Generally, neutron sputtering yields are of the order of 10 −5 atoms per neutron, except for materials containing fissionable atoms, where a yield of the order of 1 was reported. Even the sputtering yields measured in some detail, such as for Au and Nb, still give only upper limits. For very rough surfaces a small part of the material was found to be emitted as microparticles.

Experimental evaluation of the primary damage process — neutron energy effects

Experimental evaluation of the neutron energy dependence of the primary damage state depends upon a number of theoretical concepts. This state can only be observed after low- or perhaps ambient-temperature, low-fluence irradiations. The primary recoil energy spectrum, which determines the character of the displacement cascades, can be calculated if dosimetry has provided an accurate neutron spectrum. A review of experimental results relating neutron-energy effects shows that damage energy or damage energy cross section has often been a reliable correlation parameter for primary damage state experiments. However, the forthcoming emphasis on higher irradiation temperatures, more complex alloys and microstructural evolution has fostered a search for additional meaningful correlation parameters.

Calculation of radiation damage effects of 800 MeV protons in a “thin” copper target

Radiation damage effects of 800 MeV protons incident on a 1 cm thick copper target have been calculated by using the nucleon-meson transport code to determine the nuclear reactions produced by the protons, and the theory of Lindhard et al. to evaluate the resultant damage energy deposited in the target. Damage effects were found to be nearly uniform across the 1 cm target thickness, and the results obtained can be expressed in cross section form applicable to target thicknesses from about 0.01 cm to 2 cm. The calculation yielded a damage energy cross section (depending slightly on target geometry) of about 350 barn-keV, a nuclear transmutation cross section of 1.0 barn, and indicated copious proton, neutron, and helium production. Comparison of the cross sections obtained with those of several neutron spectra of reactor interest suggests that medium energy proton bombardment may be a useful method for simulating neutron-induced radiation damage effects.